Development of maintenance technique for materials of core components in light-water reactor using ion-beam irradiation

轻水堆核心部件材料离子束辐照维护技术开发

基本信息

批准号：
14380230
负责人：
HASEGAWA Akira
金额：
$ 9.28万
依托单位：
TOHOKU UNIVERSITY
依托单位国家：
日本
项目类别：
Grant-in-Aid for Scientific Research (B)
财政年份：
2002
资助国家：
日本
起止时间：
2002 至 2004
项目状态：
已结题

项目摘要

Austenitic stainless steel is used extensively in the core components of light water reactors (LWRs). Irradiation assisted stress corrosion cracking (IASCC) is one of the major concerns of stainless steel in LWR environment. Radiation induced segregation (RIS), that represents chromium depletion and nickel enrichment, caused by neutron irradiation is regarded as one of controlling factors of IASCC. To estimate the effect of chromium depletion at grain boundaries on corrosion behavior, the electrochemical potentiokinetic reactivation (EPR) test is very useful. In order to study the neutron irradiation effect of neutron, energetic particle beam irradiation using accelerators have been performed as simulation experiments. Since the corrosion behavior evaluated by EPR test is affected by chemical properties from the surface to 2-3μm depth of the austenitic stainless steel, the light ion irradiation can introduce enough damage level and damage area to evaluate the corrosion behavior of irra … More diated surface including grain boundaries. The purposes of this study are to establish the testing method using miniature size specimens for ion-irradiation experiments and to clarify the change in corrosion behavior of irradiated stainless steel caused by displacement damage under controlled irradiation conditions such as temperature and fluence.The corrosion behavior of type 304 austenitic stainless steel irradiated with hydrogen ion was investigated using EPR testing combined with TEM observation after irradiation at 300℃, 400℃ and 500℃ up to 0.1, 0.5 and 1 dpa. The results are summarized as follows ;1.It was confirmed that 3 mm disk specimens could be used to estimate corrosion behavior of austenitic stainless steel using EPR test.2.In the specimens irradiated at 300℃ and 400℃, the surface pitting was observed within grains after EPR testing. No etching trenches at and around grain boundaries were observed, nevertheless RIS was occurred in the specimens irradiated at 400℃ to 0.1dpa. So it was suggested that the corrosion resistance at grain boundaries after ion irradiation was still maintained.3.In the specimens irradiated at 500℃, etching trenches at and around grain boundaries were observed, and surface pits were absent. The width of etching trenches and the reactivation ratio increased with irradiation dose. It was suggested that high-temperature irradiation enhanced RIS and also the recovery of damage defects and therefore it suppressed the chemical activation within grains.4.The correlation between the changes in corrosion properties and the development of RIS and microstructures was discussed.Based on these results, it is concluded that hydrogen ion beam irradiation combined with EPR test is valuable to estimate neutron irradiation induced corrosion behavior of austenitic stainless steel. Less

奥氏体不锈钢主要用于轻水堆 (LWR) 的核心部件，辐照辅助应力腐蚀开裂 (IASCC) 是轻水堆环境中不锈钢的主要问题之一，即铬贫化。中子辐照引起的镍富集被认为是 IASCC 的控制因素之一。为了估计晶界处的铬贫化对腐蚀行为的影响，电化学动电位再活化（EPR）试验非常有用，为了研究中子的辐照效应，使用加速器进行高能粒子束辐照作为模拟实验，因为EPR试验评估的腐蚀行为受到中子化学性质的影响。轻离子辐照可以在奥氏体不锈钢表面2-3μm深度处引入足够的损伤程度和损伤面积来评估辐照不锈钢的腐蚀行为。本研究的目的是建立使用微型样品进行离子辐照实验的测试方法，并阐明在受控辐照条件（例如温度和注量）下辐照不锈钢因位移损伤而引起的腐蚀行为的变化。采用EPR测试结合TEM观察，研究了304型奥氏体不锈钢在300℃、400℃和300℃辐照后的氢离子辐照腐蚀行为。 500℃ 0.1、0.5 和 1 dpa 结果总结如下： 1. 证实 3 mm 圆盘试样可用于通过 EPR 试验来评估奥氏体不锈钢的腐蚀行为。 300℃和400℃，EPR测试后在晶粒内观察到表面点蚀，在晶界及其周围没有观察到蚀刻沟槽，但出现了RIS。在400℃辐照至0.1dpa的试样中，离子辐照后晶界处的耐蚀性仍然保持。3.在500℃辐照的试样中，在晶界处和晶界周围观察到腐蚀沟槽。表面凹坑不存在，蚀刻沟槽的宽度和再活化率随着辐照剂量的增加而增加。也恢复了损伤缺陷，从而抑制了晶粒内的化学活化。4.讨论了腐蚀特性的变化与RIS和微观结构发展之间的相关性。基于这些结果，得出结论：氢离子束辐照结合EPR试验对于评估中子辐照引起的奥氏体不锈钢的腐蚀行为很有价值。